This invention relates generally to apparatus and methods for operating a system and more particularly to apparatus and methods for operating a nuclear reactor.
A typical boiling water reactor (BWR) includes a pressure vessel containing a nuclear fuel core immersed in circulating coolant water that removes heat from the nuclear fuel. The water is boiled to generate steam for driving a steam turbine-generator for generating electric power. The steam is then condensed and the water is returned to the pressure vessel in a closed loop system. A plurality of piping circuits carry steam to the turbines and carry recirculated water or feed water back to the pressure vessel that contains the nuclear fuel.
The BWR includes several conventional closed-loop control systems that control various individual operations of the BWR in response to demands. For example, a control rod drive control system (CRDCS) controls the position of the control rods within the reactor core controling the rod density within the core which determines the reactivity of the core, which in turn determines the output power of the reactor core. A recirculation flow control system (RFCS) controls core flow rate, which changes the steam/water relationship in the core and can be used to change the output power of the reactor core. These two control systems work in conjunction with each other to control, at any given point in time, the output power of the reactor core. A turbine control system (TCS) controls steam flow from the BWR to the turbine based on pressure regulation or load demand.
The operation of these systems, as well as other BWR control systems, is controlled utilizing various monitoring parameters of the BWR. Some monitoring parameters include core flow and flow rate affected by the RFCS, reactor system pressure, which is the pressure of the steam discharged from the pressure vessel to the turbine that can be measured at the reactor dome or at the inlet to the turbine, neutron flux or core power, feed water temperature and flow rate, steam flow rate provided to the turbine and various status indications of the BWR systems. Many monitoring parameters are measured directly, while others, such as core thermal power, are calculated using measured parameters. Outputs from the sensors and calculated parameters are input to an emergency protection system to assure safe shutdown of the plant, isolating the reactor from the outside environment if necessary, and preventing the reactor core from overheating during any emergency event.
An essential requirement of a nuclear reactor protection system is that it must not fail when needed. Therefore, unless the operator promptly and properly identifies the cause of an abnormal transient event in the operation of the reactor, and promptly effects remedial or mitigating action, conventional nuclear reactor protection systems will automatically effect reactor trip. However, it is also essential that reactor trip be avoided when it is not desired or necessary, i.e., when there is an error in the instrumentation or when the malfunction is small enough that reactor trip is unnecessary or when one shutdown function fails, the reactor protection system must not perform the next shutdown function if to do so would be unsafe. Also, at least some known reactors include emergency cooling systems which monitor operation of the reactor.
In the event of an unsafe condition, a shut-down system or a safe operation system can automatically effect remedial action such as changing the reactor valve alignment from a normal operating mode to an emergency operating mode thereby preventing an unsafe or potentially unsafe condition. After the unsafe condition has been resolved, systems are returned to a standby mode and an operator is required to manually align the reactor for other system modes. Operator alignment may result in alignment errors and produce undesirable results. Further, the use of checklists to facilitate system alignment may require two operators to realign the system to a second operating mode.